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Rev 2 to Safety Analysis
	ML20059M316
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Site:	University of Michigan
Issue date:	08/01/1993
From:	; MICHIGAN, UNIV. OF, ANN ARBOR, MI
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{{#Wiki_filter:. .. ~ SAFETY ANALYSIS l Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 f a License Renewal Issue: 113084 Revisions: 1: 091591 2:080193 49, MICIIIGAN MEMORIAL PIIOENIX PROJECT l TIIE UNIVERSITY OF MICIIIGAN i SAFETY ANALYSIS i b I i Ford Nuclear Reactor The University of Michigan Docket 50-2 i License R-28 i 't August 1, 1993 Ann Arbor, Michigan i 9311190027 931027 f PDR ADOCK 05000002 gy P PDR 4;,

r SAFETY ANALYSIFE "I Ford Nuclear Reactor Docket 50-2,- License R l Revision 2 [ k i TABLE OF CONTENTS Page CHAPTER 1 INTRODUCTION -1 1 CHAPTER 2 SITE CHARACTERISTICS 2 2.1 Demography 2 2.2 Nearby Industrial, Transportation, and Military Facilities 2 [ P.2.1 Transportation Routes 2: 2.2.2 Nearby Facilities 2 2.2.3 Conclusion 2 t CHAPTER 3 REACTOR 3 3.1 Building Layout 3 l 3.2 Reactor Pool 8 l 3.3 Reactor Core 8 i I 3.3.1 Fuel Elements 8' 3.3.2 Reactivity Control Systems 11 3.4 Shielding 11 i i -CHAPTER 4 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 12' 4.1 Primary Coolant System 12. 4.2 Secondary Cooling System 12 ] l CHAPTER 5 -ENGINEERED SAFETY SYSTEMS 15 5.1 Ventilation System 15 5.2 Reactor Air System 15 l CHAPTER 6 INSTRUMENTATION AND CONTROL SYSTEMS 18 l 6.1 Log Count Rate System 18 6.2 Linear Level-Servo Control System 18 6.3 Safety System Period Channel C 19 6.4 Safety System 20 6.5 Additional Safety-Related Instrumentation Systems 21_ 6.5.1 Temperature System 21 1 6.5.2 Flow System 21 6.5.3 Radiation Measurement System 22 CHAPTER 7 ELECTRIC POWER SYSTEM 23 j 4 ) i i-.-

SAFETY; ANALYSIS-Ford Nucle'ar Reactor Docket 50-2, License R-28. [ Revision 2 ,j l y I CHAPTER 8 AUXILIARY SYSTEMS 24 .8.1 Fuel Handling and Storage 24 8.1.1 New Fuel 24' j 8.1.2 . Irradiated Fuel 24 8.1.3 Tools 24 i 8.2 ' Fire Protection 24 j 8.3 Heating and Ventilating 25 j CHAPTER 9 EXPERIMENT PROGRAMS AND FACILITIES 26 9.1 Experiment Programs 26 9.1.1 Neutron Activation' Analysis 26 l 9.1.2 Isotope Preparation and' } Radiochemical-Production 26

j 9.1.3 Gamma Irradiation Services 26 9.1.4

. Neutron Radiography 27 .i 9.1.5 Radiation, Chemical, and j Mechanical Testing Services 27 9.1.6 Training Programs 28 j 9.2 Experiment Faci'ities 28 l 9.'2.1 In-Core Irradiations 28 9.2.2 Pneumatic Tubes-28: -i 9.2.3 -Beamports 28 j CHAPTER 10 RADIOACTIVE WASTE MANAGEMENT -30 10.1 Solid Waste 30. 10.2 Liquid Waste-30 10.3 Airborne Radioactivity 30 -i CHAPTER 11 RADIATION PROTECTION PROGRAM 31 11.1 Personnel Monitoring and Safety 31 l 11.1.1 Personnel Monitoring 31- '{ 11.1.2 Instructions to Workers 31 i 11.1.3 Staff Exposure Guidlines 31 l 11.1.4 Visitor Exposure Guidelines 31 11.1.5 Radiation Warning. 31 11.2 Facility Internal Surveillance 32 .j 11.2.1 Floor Smear Surveys 32 11.2.2 Contamination Cleanup Standards 32 j 11.2.3 Area Radiation Surveys 32 11.2.4 Airborne Surveys 32 11.2.4.1 Mobile Ai'rborne Particulate (MAP) Monitor -- 3 2 I 11.2.4.2 Airbcrne Trifl~um Survey 32 11.2.4.3 Retention Tank Airborne Survey '32; .j 11.2.5 Pool Water Analysis 33 j 11.2.6 Hood Airflow Surveys. 33 1 11.2.7 Heavy Water Tank Tritium Analysis-33 { q 1 'l

SAFETY ANALYSIS Ford Nuclear Reactor: Docket 50-2, License R-28 H Revision 2 ~ CHAPTER 12 CONDUCT OF OPERATIONS 34 12.1-Organizational Structure and Qualifications 34 12.2 Operational Review and Audit 34 12.3 Operating Procedures 36 CHAPTER 13 BASES FOR TECHNICAL SPECIFICATIONS 37 13.1 Power and Flow Limiting Combinations in Forced Convection-37' 13.1.1 FNR Core Parameters 37 13.1.2 Heat Transfer Coefficient 38 j 13.1.3 Peak-to-Average Power Ratio 42 13.2 Power Limits in Natural Convection 47 i 13.3 Control Rod Reactivity 48 l 13.4 Core Excess Reactivity 48 13.5 Experiment Reactivity 49 CHAPTER'14 SAFETY ANALYSIS 51 14.1 Excess Reactivity Addition 51 1 14.2 Abnormal Loss-of Coolant 51-l 14.2.1 Pool Walls 52-j 14.2.2 Thermal Column 52 i 14.2.3 Piping Systems: Primary Coolant, Demineralizer, Coolant Sampling, Pool Drain, and Emergency Fill Lines 52 14.2.4 Beamports 53 14.2.5 Pneumatic Tubes 53 ri 14.2.6 Pool Drain Time and Flow Rate i Calculations 56 l 14.2.7 Emergency Makeup Water 59 .l 14.2.8 Conclusions 59 0 14.3 Failed Experiment 59 I 14.3.1 Experiment Radioactivity Limits Based Upon Exposure to Personnel Within.the Restricted Area of the FNR Building 59 j 14.3.1.1 Assumptions 59 l 14.3.1.2 Calculations 60 l 14.3.2 Experiment Radioactivity Limits Based Upon Exposure to Personnel in Unrestricted Areas 60-14.3.2.1 Assumptions 60. 14.3.2.2 Calculations 61 14.3.3 Limits on Single and Double Encapsulation Experiments 62 14.3.4 Fissile Material Experiment' Activity Limit 62 i 14.3.4.1 Calculations-- 63 l ) i l 9 'l I

l SAFETY' ANALYSIS Ford Nuclear Reactor Docket-50-2, License R-28 g Revision 2 ? l 1. INTFODUCTION i This safety analysis supports the operating-license renewal i application for the Ford Nuclear Reactor (FNR)..The analysis provides a description of reactor site, building, pool and core arrangement, mechanical and instrument systems, experiment programs and facilities, radiological waste management, radiation I protection program, conduct of operations, and accident analysis. Additional specific data concerning the reactor, fuel, heat transfer, core neutronics, operations, and laboratory and experiment facil'ities is available in the Research, Training, Test, and Production Reactor Directory, Second Edition, 1983, pp. 467-474, published by the American Nuclear Society. r ? 4 s I i i f B I 1 ? s._ i i =

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 . Revision 2-2. SITE CHARACTERISTICS l i The Ford Nuclear Reactor'(FNR) is located on the North i Campus of the University of Michigan at Ann Arbor, Michigan. The l North Campus area is.under the administrative control of the Regents of the University of Michigan. The North Campus is a tract of nearly 900 acres, approximately 1.5 miles northeast of l the center of Ann Arbor. It is bounded on the north by Plymouth Road and on the so"th by Glacier Way. Open land and the Arborcrest Cemetery lie to the east. To the west are University athletic fields, municipal parks, and a wooded ridge. The Huron River i flows through land bordering the area on the west and south and i some marshland lies adjacent to the river on the south. 2.1 Demography The reactor building and the contiguous Phoenix Memorial l Laboratory (PML) are located. ar the center of the North Campus ~ area. No housing or bu_ildings containing housing facilities 2 **. ? are erected within '500 feet of the_ reactor. Reference to Buildings' 2.2 Nearby Industrial, Transportation, and Military Facilities ithin 500 w feet deleted. I 2.2.1 Transportation Routes .i Transportation routes located-close to campus include Plymouth Road which is located approximately 2500 feet l north of the reactor site, and Glacier.Way which is located 1500 feet south of the reactor site. There are no main railroad lines near the campus. 1 1 2.2.2 Nearby Facilities There are no heavy industries or major military establishmente in the vicinity of the North Campus. 2.2.3 Conclusion There is no heavy air or railroad traffic, no heavy truck route, and no heavy industry close'enough to the North Campus to constitute a threat to safe operation of the reactol. .l 1 M- ?i I 2

. - ~... - ~ i i SAFETY ANALYSIS i Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 3. REACTOR i The Ford Nuclear Reactor is a two megawatt, open pool l reactor facility. The heterogenous core is composed of aluminum and enriched U-235. It is suspended 20 feet beneath the surface l of the pool from a moveable bridge which is mounted on rails that lie on top of the concrete tank. The reactor operates at a l licensed power level of two megawatts. The FNR generates no electricity and is used by students, faculty, and staff of the University of Michigan for research, experiments, and classes. In I addition, services are provided to other universities, colleges, schools, and institutions and to industrial research j organizations. Principal utilization includes neutron irradiation services, neutron activation analysis, isotope preparation, radiochemical production, gamma irradiation services, neutron radiography, testing services, and training programs. i I 3.1 Building Layout (Figures 3.1, 3.2, 3.3, and 3.4) i The reactor building is a windowless, four story, reinforced I concrete building with 12 inch walls structurally integral with i the footings in foundation mats. The building is approximately 69 l feet wide by 68 feet long by 70 feet high with approximately 44 i I feet exposed above grade. The building has the following general features:

1) the reactor is housed in a closed room designed to restrict leakage; 2) the reactor room is equipped with a l

Ventilation system designed to exhaust air or other gases present + in the building atmosphere into a stack above the cooling tower which exhausts a minimum of 54 feet above ground level; 3) the ventilation system provides ventilation for certain storage and l experimental facilities and exhausts these a minimum of 54 feet above ground level; 4) the openings into the reactor building are an equipment access door, three personnel doors, an equipment i access hatch, air intake and exhaust ducts, room 3103 fume hood exhaust duct, a beamport ventilation duct, a sealed north wall .i door, a door between the hot cave operating face and the beamport 1 floor, a sealed foundation tile drain to the cold sump, and a l2 1 pneumatic tube system for sampie transfer between the FNR and several laboratories in the PML. l t t 6 3

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J SAFETY ANALYSIS 'I Ford Nuclear-Reactor Docket 50-2, License R-28' Revision 2 q 3.2 Reactor-pool-(Figure 3.5) The rectangular reactor pool is 27 feet deep andLis constructed of barytes concrete to a height of 15 feet, the remainder being ordinary concrete. The tank is approximately 10 feet wide by 20-feet long and contains approximately 50,000 gallons of demineralized water. The pool lining consists of white ceramic tile sealed with white cement. The tile protects the concrete from spalling, aids visibility, and is more easily [ decontaminated than a concrete surface. Spent fuel is stored in racks along the walls of the reactor pool. Storage areas in.the-pool are for depleted fuel which is being prepared for shipment to a reprocessing facility and for partially depleted fuel which can l, be reused in the reactor core. A water lock system in the. south j end of the pool provides a means for transferring highly radioactive samples, experiments, and reactor fuel from the pool-i to a shielded hot cave. Experiment facilities for high power-operation include beamports,' pneumatic tube stations, and a large space within the pool for sample irradiations. j 3.3 Reactor Core (Figure 3.6) The reactor core is suspended 20 feet beneath the surface of 'the pool from a moveable bridge which is mounted'on rails that lie. i on top of the concrete tank. A typical core. configuration consists of 35-40, 19.5% enriched, MTR plate type fuel-elements. I Standard elements contain 167 gram of U-235 in 18 aluminum clad fuel plates. Control elements, which have control rod guide. j channels, have nine plates and contain 84 gram of-U-235..Overall fuel element dimensions are approximate 1y 3 inches x 3 inches x 26 inches. The reactor produces a peak thermal flux of approximately n/cm /sec. The reactor core contains three shim rods [2 l 2x1013 2 and one regulating rod. A heavy water tank is located against the north face of the reactor core. The heavy water is used to provide a thermal neutron flux for beamports. 3.3.1 Fuel Elements ( The fuel elements are MTR plate type. Each of the plates is nominally 3 inches x 25 inches x 0.06 inches with-an active fuel width and length of 2.6 inches x'24 inches. plates are fabricated in a sandwich fashion with 0.015. inch aluminum cladding on each side of a 0.030 inch thick layer of'42 Wt% uranium in uranium aluminide-aluminum. The fuel' meat contains approximately 9.3 g of U-235. l u-l l 8 e a

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SAL TY ANALYSIS. ) Ford Nuclear Reactor' Docket 50-2, License R-28 . Revision 2 i.c Figure 3.6 Reactor Core t Horizontal I F g Neutron 'i P.adiography Beamport Beamport. Facility E H Beamport Beamport D Power Linear Power Beamport Log N. Ran e B Level Range A I Beamport O ) i 'i Heavy Water Tank r I 1 i 5 Vertical Neutron Radiography Facility l r Bea port 65 55 45 35 25 15. Beamport - A C { 76 66 56 1 1 36-I 16 6~ Rod Rod 77 67 57 47 37 27 17. 7 { REACIOR-b watroA CORE 78 68 58 I I 38 I I 18 8 g 3 Irradiation 79 69-59 49 39 29 19 9 System O80 Fission Chamber 2 Fission Chamber 1 1 70 60 50 40 30 20 10 i A Aa l ? n.; 10 1 ,I

i SAFETY' ANALYSIS l Ford Nuclear Reactor i Docket 50-2, License R-28 Revision 2 .j 3.3.2 Reactivity Control Systems Reactivity control of the FNR is provided by three boron stainless steel shim safety rods and one stainless steel control rod. All rods move in vertical slots within control fuel elements. l Shim safety rods are driven through gear reductions-l by motors located beneath the reactor bridge structure. The shim j safety rods are attached to rod drive extensions by magnets. The. i shim safety rods are the primary means for shutting down the reactor. Upon receiving a scram signal, the magnets are deenergized and all three shim rods drop by gravity. i l The control rod drive motor is located on the bridge structure. The control rod is attached directly to its rod drive extension. There are two modes of operation for the control rod, manual and automatic. In manual speration, rod motion is controlled by a raise-lower switch on the reactor control console. In the automatic mode, the control rod is automatically adjusted-in response to the signal from the linear level system compensated ion chamber to regulate reactor power at the desired setpoint. The shim rods have a total reactivity of approximately 0.06 delta K/K. Each of the three rods is of nearly l equal worth. The control rod worth is approximately 0.003 delta K/K. Shim rods are normally controlled as a group. Shim rod l speed is 0.044 in/sec. Control rod speed is 0.533 in/sec. 3.4 Shielding Biological shielding is provided around the FNR to minimize i the exposure to any individual working within the laboratory to levels a, low as reasonably achievable. The' biological shielding l is made of barytes concrete to a height of 15 feet, the upper level of the reactor tank is standard structural concrete. The walls of the lower section are 6} feet thick. Biological shielding is provided above the reactor core by the pool water. i i I M* t 11 i

SAFETY-ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 4. REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 4.1 primary Coolant System (Figure 4.1) The components of the primary coolant system are the header I and hopper mechanism, the hold up tank, a pump driven by a 20, l2 hp electric motor, the heat exchanger, and associated piping and .l instrumentation. The primary coolant system removes 2 Mw of heat ^ from the core by forced circulation and maintains the bulk pool temperature at less than 116 F. The flow rate is between 900 and 1000 gpm. A moveable header is positioned beneath the reactor l core to provide forced circulation when the reactor is operated in i the forced coolng' mode. The header is attached via a rotating l flange to primary coolant piping in the pool floor. In the forced circulation mode of operation, primary coolant flows down through the core fuel elements, grid plate, hopper, and into the header. From the header it passes into the holdup tank and to the primary pump where it is pumped to the heat exchanger. From the heat exchanger, primary coolant flows through the primary flow orifice and returns to the pool. Approximately 25 gpm is tapped off j e downstream of the flow orifice and flows to the primary demineralizer system (hot DI). i 4.2 Secondary Coolant System (Figure 4.2) The heat produced during operation'of the reactor is dissipated to the atmosphere by the secondary coolant system. l Secondary coolant absorbs heat from primary coolant in a cross flow heat exchanger. The secondary water, in turn, is cooled by evaporative cooling in cooling towers. The secondary system operates at atmospheric pressure and is driven by a single, 25 hp, I centrifugal pump that produces a flow rate of approximately 1,000 gpm. 1 l i

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i SAFETY ANALYSIS f Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 I f 5. ENGINEERED SAFETY SYSTEMS l The only engineered safety systems relate to the ventilation i system for minimization of effluents released to the atmosphere 'I and the reactor air system as utilized for component operation and j control. j 5.1 Ventilation System (Figure 5.1) .I t f The heart of the ventilation system is the FNR fan room, room 2111, where all supply air enters the building and where the predominent amount of exhaust air exits the building. The supply i intake and the building exhaust have isolation dampers which act l from a common pneumatic cylinder. They are opened and closed simultaneously Supply air enters the reactor building through l the supply damper, passes through a set of filters and into the i main supply blower for distribution throughout the building. i t The predominant amount of reactor building air is exhausted ( through the main reactor buildir.g enbr.ust damper. Air from the

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beamport floor, pneumatic tube system blc wer, and a hood in room l 3103 exhausts into PML stack #2. A high radiation level of greater -l than one mrem /hr in the reactor building exhaust air activates the j building radiation alarm that scrams the reactor, shuts all of the ^ supply and exhaust dampers, and turns off the supply and exhaust fans. A high radiation level of 5 mrem /hr in the reactor fuel i vault initiates the same sequence of events. ] Building supply and exhaust systems are continuously l monitored for radioactivity. Stack #2 has a mobile air particulate monitor (map) and a gaseous activity detector (GAD). The reactor building exhaust has a continuous gaseous activity i detector. The pool floor and the beamport floor are monitored { continuously by MAP monitors. PML exhaust stacks 1, 2, 3, and'4 have isokinetic filters through which continuous air samples are taken for periodic monitoring. In addition, the same sample lines j have a charcoal filter to monitor for iodine. The reactor t building exhaust has similar isokinetic particulate and charcoal j iodine monitoring systems. i 5. 7. Reactor Sem* ice Air System The reactor service air system has two reactor air 1 compressors located in room 2077. The main. compressor is reactor air pressure of _7jl psig, charges the 2 activated at a reactor air reservoir, and cuts out at a pressure of 80 psig. A backup compressor actuates at approMimately Gj[ psig in the event of failure of the main compressor. From room 2077, reactor air is distributed for use in the reactor building and in the Phoenix Memorial Laboratory. On the pool floor, reactor service 15 1

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 l 4 Figure 5.1 PML-FNR Ventilation System Phoenix Memorial Laboratory Code: B = Blower C = Charcoal Filter GAD = Gaseous Activity Detector I = Isokinetic Filter l MAP = Mobile Air Particulate Monitor e P = Air Pump RM = Radiation Monitor l l I I I I Exhaust 4 3 2 1 Stacks fiAP P GAL Rm 1069 Ford Nuclear Reactor X X Pool Floor Pm 3103 MAT MAP Hood l O Pool Floor Fn 2105 Fuel Second C vault -4 EM I o Orlwr Pm 2102 5 mrem /hr Beamport and Storage Port g B ea.raport C O C g Floor Vent System h Q " lower

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SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2,. License R-28 l Revision 2

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i l air is-available at four outlets ecound the perimeter of.the reactor pool. Service air is also available in room 2103. Service air is utilized in the control room for cooling. tower sump' 1 l level control. In room 2111, service air is utilized for control I of the reactor building supply and exhaust ventilation dampers, i the supply and exhaust ventilation duct bypass, and for pressurization of the building hot water system. On the.beamport floor, service air is utilized at J-beamport, in the cooling tower sprinkling system, for steam valve control, and as service air in the vicinity of the radioactive storage ports. In the reactor basement, service air is utilized at the hot and cold demineralizers for resin separation and mixing'and with the i secondary water makeup valve control system. In the Phoenix Laboratory, in room 1033, reactor air serves as a backup-for the PML heating and ventilation system air compressor. i i i .i ~! t ') a p + i 17

.. = - SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision-2 6. INSTRUMENTATION AND CONTROL SYSTEMS [ Six neutron detection channels measure reactor power from the source range to full power: (1) two fission chamber channels capable of accurately measuring power from source range through full power; (2) two compensated ion chamber channels capable of i p measuring power from the intermediate range through full power, and (3) two uncompensated ion chamber, power range channels I capable of accurately measuring power between 500 kw and 2 Mw. High power scrams are initiated by each of the power range j channels at 2.4 Mw. A five second period scram is initiated by one of the intermediate range, compensated ion chamber channels. } t In addition to power level channels, the instrumentation and i control systems provide control, alarms, and indications for temperature, flow, radiation level, control rod position, and reactor pool level. 61 Log Count Rate System l There are two identical channels, each consisting of a fission chamber, a preamplifier, a linear amplifier, an electronic scaler, a pulse height discriminator circuit, a count rate meter, j and a count rate recorder. Fission chambers are positioned on the l cast and west side of the core. At startup, each fission chamber I is placed near the reactor core. Fission events within the f chamber produce current pulses which are converted to voltage pulses, amplified, shaped, and then counted on the scaler. Smaller pulses, produced by other radiation are rejected by the j pulse height discriminator circuit. These amplified pulses are also fed to a log count rate meter where they are integrated, and their average counting rate is displayed on a logarithmic indicating meter calibrated for five decades (1-100,000 cps). The average counting rate is recorded on a recorder provided with five cycle logarithmic paper. As the power level of the reactor is increased and the counting rate increases, the counting rate channel reaches its upper limit, at which time it is repositioned to maintain on scale indication. During reactor startup, the Log Count Rate recorders are interlocked to prevent shim safety rod motion unless indication is greater than 5 cps. i 6.2 Linear Level - Servo Control System i The linear level-servo control _ system consists of a compensated ionization chamber and its power supply; a range i flux l2 selector; a recorder; and an automatic controller. Because this is the most precise channel for measuring the neutron level, it is used for automatic control of reactor power level. [ l i e i 18 i ..C',

I SAFETY ANALYSIS" l Ford Nuclear Reactor i Docket 50-2,. License R-28 Revision 2 i i The compensated ionization. chamber provides an output current proportional to the neutron flux at the chamber. The 1 output of the chamber is fed to a range selector which, in turn, j provides a suitable input signal to the recorder and to the j2 controller. The controller signals a servo control unit,-which, through a drive unit, gives the control rod drive the necessary impulse in a given direction to maintain the power level at the control setpoint. If power level is above the.setpoint, the l control rod is inserted into the reactor so that reactivity is l decreased and power level is restored to the control point i setting. Conversely, whenever the power level drops below the control point setting, the control rod-is withdrawn so that j reactivity is increased and power level increases to the control j point setting. During reactor startup and power level changes, the range selector switch must be periodically repositioned to keep indicated power on scale. The recorder has a switch which, if activated at 2.3 Mw, inserts the three shim safety rods at their normal speed. This interlock, called an auto rundown, acts before the scram setpoint of 2.4 Mw. Further, an alarm circuit connected i directly to the linear level system ion chamber warns the-operator j -when power reaches 2.2 Mw. -t The servo control system drops out of automatic control when i indicated power is below the setpoint by more than 5%, the range j selector switch is repositioned, or the control rod is manually l repositioned. Should the servo system malfunction and remain in l automatic control, the control rod is inserted if indicated power drops 15% below the setpoint This interlock prevents automatic rod withdrawal which would be the normal sequence if indicated power were below the setpoint. If indicated power is above the 1 setpoint, there are no interlocks because control rod insertion, I which is the normal command sequence,. acts to safely-reduce power. 6.3 Safety System period Channel C l2 l The Safety System period Channel C detector is a l2 l compensated ion chamber located in the reactor pool behind the r ) north end of the heavy water tank. The ion chamber receives high voltage and negative compensating voltage from the Channel C [2 module located in the control room instrument panel. The output signal from the ion chamber is transmitted to the Channel C l2 module. The amplified detector signal is output from the Log N recorder. The power level signal is converted to a period signal within the Channel C module. The resultant period is output to the period recorder. { l t i i k 19 w ~ e.e r w

_~ l SAFETY ANAhYSIS' .) Ford Nuclear Reactor l Docket 50-2, License R-28 i Revision 2 j f At 5% power or 100 kw, a_ switch within the Log N recorder i provides a contact closure signal for both the high power /no flow and high' power / header down scrams. The period recorder has two i inhibit switches associated with.it. A 30 second inhibit prevents -i the control rod from being withdrawn if reactor period is less than 30 seconds. Should period reach less than 10 seconds, an ] automatic rundown of-the shim-safety rods is initiated. 6.4 Safety System i The reactor safety system is designed to shut the reactor down and maintain a safe, shutdown condition. The cafety-system 3 consists of three separate channels, any'one of which can act to' t scram the reactor. Channels A and B monitor reactor power. Each l will separately initiate a scram at 2.4 megawatts. Channel C i monitors reactor period and initiates a scram-for a period o seconds or less. Reactor power level' signals are transmitted to Channels A and B from uncompensated ion chambers positioned behind the i reactor heavy water tank. Channel C receives its power signal from a compensated ion chamber in the same location. The Channel C power-level signal is converted _to period.within.the. channel C module. The ion chambers are powered by chamber' power supplies i housed in each module. .If the scram setpoint is exceeded in any channel, power 2 r to the magnet power supply is turned off and the current to all three magnets is turned off. In addition to reactor scrams caused by high. power level and short period, the foll'owing 'I actions scram the reactor by deenergizing the 115 vac power to'the magnet power supply: Scram Interloch Interlock Function Header up-zero flow Prevent reactor operation above 100 kw without forced circulation. Bridge not clamped Prevent core movement during .f power operation. l2 l High power-header down Ensure forced circulation for' power operation. Shim Range l Defeat i Scram Deleted ~ t i 20 s 1

i _ SAFETY ANALYSIS' i _ Ford Nuclear Reactor Docket 50-2. License R-28 -l Revision 2 l 1 i a l Beamport door open Prevent ~ personnel radiation j exposure at power. -l2 High power-no flow Ensure forced circulation for l power operation. . Thennal ' colunm door i Radioactivity in >l mrem /hr at FNR building sermn deleto ' -{' building air exhaust; >5 mrem /hr in FNR fuel vault; Provide indication of fuel' failure or vault criticality. .[ The time required for the shim rods to be fully inserted' following initiation of a scram is less than 0.5 seconds. i 6.5 Additional Safety Related Instrumentation Systems 1 In addition to instrumention dedicated to safe reactor I operation, the following safety related systems are also i installed. } 6.5.1 Temperature Systems The temperature system consists of a temperature recorder and temperature detectors. The temperature detectors are-t 100 ohm platinum resistance devices, familiarly described as thermohms. Each device's resistance varies with temperature. Each thermohm serves as one leg of a Whetstone resistance bridge that is utilized to generate a measured signal. j i The temperature recorder'is a single pen' recorder-j that cycles through and records multiple input temperatures. 7 Primary coolant (bulk pool and heat exchanger, inlet and outlet) l and secondary coolant (heat exchanger inlet and outlet)'are-monitored. The recorder has a high temperature auto rundown set i for 129 F. Any primary coolant temperature that exceeds 129 F 2 will initiate a rundown. ] 6.5.2 Flow System i Two primary differential pressure transducers sense f in parallel across the primary coolant flow orifice. A single. i secondary transducer senses across the secondary coolant flow .i orifice. The transducers are strain gauge devices that transmit a j current signal proportional to the pressure differential across the strain gauge. 21

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2 License R Revision 2 I i Each flow signal is transmitted to a square root extractor, since flow is proportional to the square root of differential pressure. The output voltages of the square root extractors are sent to digital flow indicators, and the primary i signals also go to alarm modules. Primary low flow alarms are l activated if flow is less than 900 gpm. The alarms provide no flow signals for the high power-no flow and header up-no flow scrams. 6.5.3 Radiation Measurement Systems The radiation measurement system consists of: (1) mobile air particulate (MAP) monitors which monitor airborne particulate activity on the pool floor, on the beamport floor, and at stack #2; (2) gaseous activity detectors (GAD) which monitor gaseous activity in the exhaust air from the reactor building and at stack #2; and (3) local area monitors (LAM) which measure gamma [ radiation levels throughout the reactor facility. 1 LAMS for the building air exhaust, reactor bridge, j north wall of the beamport floor, northwest column of the beamport floor, and the northeast column of the beamport floor, as well l2 as the GAD for the reactor building exhaust air, all feed into the radiation recorder on the control room instrument panel. If the $$d{ radiation levels from any of these inputs to the recorder exceed Demineralizt _their setpoints, the high radiation recorder alarm annunciates l2 UW Nieted i on the control room console. When radiation levels in the building air exhaust plenum exceed 1 mrem /hr, a building air radioactivity alarm sounds on the control room conscle and the t building alarm is actuated. This signal scrams the reactor automatically, shuts down the ventilating fans, and closes the reactor building supply and exhaust dampers. The fuel vault monitor has a local alarm at 5 mrem /hr. That signal also actuates the building alarm and performs the identical functions of the i L building air exhaust plenum monitor alarm. An area monitor i located in the room 3104 hood exhaust line as well as the pool floor MAP and the beamport floor MAP have local alarms. The GAD and MAP which monitor stack #2 have local alarms and send a remote I stack alarm signa] to the control console. t I r M~ 22

.. ~ .I SAFETY ANALYSIS Ford Nuclear Keactor Docket 50-2, License R-28 Revision 2 j j 7 ELECTRIC POWER SYSTEM Primary electric distribution to the FNR-FML facility is a 480 volt power line which enters room 1033. This primary supply line is bus tied to similar lines in two nearby buildings. 480, 240, and 120 vac is distributed from this power distribution center to the reactor building. Separata. circuit breakers are provided for reactor control systems and for reactor building service. i In the event of loss of normal reactor power, emergency power is supplied to essential loads by the emergency generator. In general, those loads can be classified as radiation monitoring s equipment, a limited amount of building lighting, alarm systems, the FNR ventilation system, and auxiliary equipment including the l bridge drive circuit, the heavy water tank control circuit, the i telephone system, the pneumatic tube air blower system, and the backup reactor air compressor. The emergency generator is in room 2077. The bus transfer, which switches from the normal building .I' supply to the emergency generator, is located in room 2074. The j emergency distribution panel is located on the beamport floor. Reactor control power from room 1033 is transmitted to filters in room 2111. The filter outputs provide basic 120 volt l ac power for the reactor control system. A vertical instrument i panel in the control room is the center of reactor control power distribution. Signal leads and detector leads from the reactor l bridge pass through conduit down into room 2111. Detector leads i are routed via the east side of the pool. Signal leads are routed. via the west side of the pool. Conduit runs from room 2111 are routed in the bottom of the instrument panel. All signals from J instruments in the basement are routed via cable junction boxes in the basement, room 2109, and room 2111 into the bottom of the instrument panel. Incoming signals required to go to the control i console in the control room are routed from the instrument panel. l I e ? 23

~ SAFETY' ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 I 8. AUXILIARY SYSTEMS J 8.1 Fuel Handling and Storage 8.1.1 New Fuel i New fuel is handled so as to prevent mechanical .l damage or soiling by hands or other materials. Fuel is stored within the fuel. vault and in the pool in geometric arrays that assure suberiticality and that permit sufficient natural I circulation cooling by water or air such that fuel element temperature will not exceed 100 C. As a temporary condition while transferring. unirradiated fuel between the vault and the reactor pool, the fuel i can be laid down in a row on a flat surface with not less than 6 inches between each element. Before putting a' fuel element into -l the pool, it is inspected for damage and for blockage of-coolant l channels. i 8.1.2 Irradiated Fuel In general, fuel stora ;e racks in the south end of the pool are utilized for depleted tuel being prepared for -l shipment,.and fuel storage racks in the north end of the pool are utilized for partially depleted fuel which can be reused in the reactor core. 8.1.3 Tools Two types of aluminum tools are utilized for j handling fuel elements. The fuel handling tool.for standard fuel l elements is a 26 foot long aluminum rod with a nose hook on the submerged end. The tool is hooked over a bail on the top of the i standard element, latched, and the element can be moved within the i pool. ' Control rod fuel elements do not have the fuel handling j bail because of the guide slot for rods. On a control element, a the top of the element is slotted. The control. fuel element 2 handling tool ~has.a hook which.is placed through the control' element slot for handling. 8.2 Fire Protection j n. Since the construction materials of the reactor, such as concrete walls,. brick, and floor tile, are predominately i nonflammable, a serious fire is considered to be very unlikely. 1 The fire alarm is manually activated'When a fire exists which is-more-than minor. Sounding of this alarm calls for a general evacuation of the FNR-PML facility. [ l 1 .l 24 i

i SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2,-License R-28 i Revision _2 l t The. fire alarm bell is automatically initiated by three smoke detectors located in room 2054 (the PML library), room 3051, j and room 3021. It is also initiated by the cooling tower sprinkler system. The fire bell can be initiated manually at any fire alarm pull box. All fire alarms are transmitted via telephone lines to the University Department of Safety and Public Security. The officer on duty contacts the FNR-PML i emergency staff and the fire department. l ? 8.3 Heating and Ventilating q Steam for heating the reactor building is supplied from boilers in the adjacent Cooley Laboratory, All areas are ventilated by a forced, warm air system. j i f I f I f i I I f f i I l +- i 25

f SAFETY ANALYSIS Ford Nuclear Reactor l Docket 50-2,. License R-28 Revision 2 9. EXPERIMENT PROGRAMS AND FACILITIES i 9.1 Experiment Programs 9.1.1 Neutron Activation Analysis The highly sensitive analytical technique of. neutron I activation analysis is available as a service performed by the s laberatory staff or to be performed directly by researchers using the laboratory's facilities. Neutron activation analysis is a method of identifying and measuring minute quantities of certain trace j elements in many types of materials. Approximately 39 common -l elements become' radioactive when exposed to the neutron flux in t the reactor. The subsequent radiation produced by the decay of the activated nuclei is characteristic for each element and i permits identification. The technique is particularly useful for' i analyzing environmental samples and to analyze industrial samples [ for mair taining quality control. The sensitivity, accuracy, variety ..f types of materials that can be analyzed, large number of elements that can be detected, and essentially non-destructive j 3 nature of the technique make neutron activation analysis an excellent analytical tool. l I 9.1.2 Isotope Preparation and Radiochemical Production j I Preparation of and custom labelling with radioisotopes is available for medical and industrial research, j Five radioisotopes and radiolabelled chemicals are routinely j produced by the laboratory. Elemental bromine-82 is produced for pharmacological research. Bromine-82 labelled motor oil is j prepared for use in research programs to help improve engine. oil economy. Fluorine-18 in saline solution is used for tumor . localization studies in bone. NP-59, an iodine based investigational drug approved by the Food and Drug Administration. is produced in large quantities for use in the. diagnosis of i adrenal gland diseases. Nearly 100' hospitals in the. United - l States,. Puerto Rico, Canada, and' Scotland receive regular .. j shipments of NP-59 from the facility. Sodium-24'is supplied bv J the laboratory to the University's Physiology Department for brain capillary permeability studies. 9.*.3 Gamma Irradiation Services j i A cobalt-60 source of.approximately 25,000 curies [2 is available for gamma i rradiations. '" Typical applications include sterilization of bones and cartilage for human grafts, i

z sterilization of animal food for germ-free animal colonies, j

radiation pasturization of food, studies of radiation effects on ~ 4 26 4

- _ ~. _ _.. i SAFETY ANALYSIS Ford Nuclear Reactor-Docket 50-2, License R-28 Revision 2 I chemical systems, electronic components, biological material, j animal populations, and crystals, and irradiation of seeds and l plants to change growth and develop mutants. The peak dose rate in the center well of the mobalt-60 source is approximately 1x106 l rad / hour. Gamma irradiations can also be performed in the reactor i spent fuel storage racks where the peak dose rate is approximately 3 1x105 rad / hour. The fuel storage racks are particularly useful for irradiating large objects. 9.1.4 Neutron Radiography t Neutron radiography services are available to researchers who wish to pursue problems in non-destructive testing. Neutron radiography is a technique similar to X-ray radiography except that neutrons, unlike X-rays, interact with j atomic nuclei rather than outer electrons. Whereas dense materials such as lead, iron and uranium are opaque to X-rays, they are easily penetrated and examined with neutrons. Neutron radiography also reverses to an extent the relative order of imaging possibilities. For example, details of plastics, oil, [ water, and fractures or voids inside heavy materials can be determined with good resolution. j The facility currently operates two neutron i radiography facilities. A three inch diameter facility with a .i length to diameter ratio of 300 provides extremely fine resolution } of small objects. A larger facility associated with a beamport I can produce full eight by ten inch radiographs with excellent resolution. The facility has a length to diameter ratio of 50 and neutron intensity variations of not more than ten percent over the l exposed film. i 9.1.5 Radiation, Chemical, and Mechanical Testing Services-6 Complete materials testing programs can be conducted at the laboratory. Neutron and gamma' radiation damage studies-can { i be performed in the reactor core, in spent fuel storage, and in i the cobalt-60 source, Neutron attenution tests through shieldinJ l materials are performed utilizing beamport spectrometers and ~ neutron radiography. Gamma attenuation tests are' performed with i small, well collimated gamma sources. i Mechanical and chemical tests include tensile i strength, cantilover flexure, dimensional stability, weight changes, specific gravity, hardness, and gas evolution and analysis. t i t k .i .7 27

I SAFETY ANALYSIS I Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 i i 1 l i 9.1.6 Training programs 1 Training programs are offered in neutron activation j analysis and reactor operations and instrumentation. j One and two week reactor operator and instrument l technician training sessions have been developed for electric i utility companies. These sessions are combinations of lectures, problem sessions, reactor experiments, and operational training. 9.2 Experiment Facilities (Figure 9.1) The Ford Nuclear Reactor provides three methods of neutron-irradiation: in-core, pneumatic tubes, and beamports. l r 9.2.1 In-core Irradiations ~ Targets of various sizes and shapes can be irradiated by placing them in or near the reactor core. -At 2 Mw' thermal neutron flux level at the center of the core is n/cm /see and at the edge of the core is l2 approximately 2x1013 2 n/cm /sec. Sample irradiations can be l 2 approximately 1x10L3 conducted for periods as sh~ ort as a few minutes and for as long as a year or more. j 9.2.2 pneumatic Tubes pneumatic tubes can be used to irradiate small l2 targets for up to two hours. Minimum irradiation time is one second. The thermal neutron flux in the pneumatic tubes ranges 2*NfL { n/cm /see to 4x1012 n/cm2 /sec. Targets up to 15/16 P-tubes - 2 from 2x1012 inch in diameter and 2 inches long can be irradiated in deleted j polyethylene containers. i 9.2.3 Beamports l i Ten horizontal beamports can be utilized for.long term irradiations and neutron beam extraction experiments such as j neutron radiography and neutron spectroscopy. The beamports j penetrate the reactor pool' wall and terminate at a heavy water j tank-adjacent to one face of the core. The heavy water tank provides a cadmium ratio of approximately 11. Thermal neutron j flux from the beamports varies between 1x108 n/cm2 /sec and 1x107 .i -n/cm /sec. l 2 i I J 28 y

SAFETY ANALYSIS Ford Nuclear Reactor-I Docket 50-2, License R-28 Revision 2' I Figure 9.1 Ford Nuclear Reactor Experiment Facilities r i i ~i - i r G i j @x x @./ } l ' N l\\ O O o'o o y / e% O RsA o O l ] / O D 'N b O -\\, l / MS$$MN MSM2M O 4 iG#MS#M 4 'N

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SAFETY ANALYSIS Ford: Nuclear Reactor Docket 50-2, License R-28 l Revision'2 l 10. RADIOACTIVE WASTE MANAGEMENT i 10.1 Solid Waste i Solid radioactive waste is collected and stored in 55-gallon steel barrels. If waste materials are to be l2 [ discarded the health physicist provides assistance in transferring materials to designated storage areas. The waste is placed in 1 containers for storage or shipment. In temporary situations, l2 radioactive waste materials may be stored in work areas. Radioactive waste containers that are to be shipped from FNR-pML facility to an authorized disposal site comply with~ Department of Transportation (DOT) regulations. ] f 10.2 Liquid Waste Radioactive liquid wastes are collected in the facility hot sump and cold sump. Liquids from the sumps are pumped to one of l: three radioactive liquid retention tanks. Each tank has a capacity of 3000 gallons. Weste-water is normally filtered, 2 demineralized, and recycled as make-up to the reactor pooi. Nuclear Regulatory Commission regulations, 10CFR20, permit the University of Michigan to discharge 1' curie pe. year of i soluble radioactive liquids other than tritium and carbon-14 l2 to the sanitary sewer system. The FNR-pML is alloted 1/2 curie [ per year by the University. Normally, the facility does not 2 l discharge radioactive li qu ir's to the sanitary sewer. 10.3 Airborne Radioactivity The PML exhaust stacks and FNR building exhaust are monitored continuously for airborne particulates. A particulate [ activity record is maintained. } The FNR building exhaust, FNR beamport and hood exhaust, and PML hot cave exhaust are continuously monitored to detect argon-41 produced in'the reactor pool, beamports, and pneumatic tube- .i system. The PML exhaust stack and FNR building exhaust are i monitored continuously and analyzed to detect fission product A iodine-131 leakage and radioisotope iodine discharge. Selected areas surrounding the FNR-FML facility on the North-Campus are monitored for particulate activity, airborne, iodine, [ and whole body exposure rates as_part of a routine, environmental ~~ sampling program. e ? 30

._ ~ SAFETY ANALYSIS' ] Ford Nuclear: Reactor. ] Docket 50-2. License R-28 . Revision 2 .i .11. RADIATION PROTECTION PROGRAM { 11.1 Personnel Monitoring and Safety ? 11.1.1 Personnel Monitoring i All personnel entering the FNR-PML facility are issued film badges except: (1) visitors and tours who'are under constant escort; and (2) individuals who will be working in l ^ non-radiation areas, such as. meter-readers and copy machine-maintenance personnel, specifically authorized by1the Reactor j Manager, Assistant Reactor Manager, Radiation Laboratory Manager, or health physicist. '[ 'i Film badges are worn in. radiation and high radiation areas. Finger badges or tabs are issued to personnel who work in intense radiation beams, and are worn during the period of such j work. Both film badges and finger tabs are developed monthly. 11.1.2 Instructions to Workers 'i Personnel entering the FNR-PML facility other than r visitors and tours receive instructions in the areas prescribed by j 10CFR19.12. All employees and experimenters receive health physics instructions and, in addition, female employees and j I experimenters receive instructions on the possible health risk to children of women who are exposed to radiation during pregnancy. i 11.1.3 Staff Exposure Guidelines i Every effort is made to keep exposures as low as .i reasonably achievable. During normal operations,.a staff member [ is not to allow himself or others to be exposed to greater than ? 100 mrem /wk whole body and I rem /wk to the extremities. j 11.1.4 Visitor Exposure Guidelines Every effort is'made to minimize visitor and tour exposures. Tours are not permitted in high or very high 2. radiation areas. Visitors are permitted in high or very high radiation areas only with specific approval of reactor ' management or the health-physicist. } i 11.1.5 Radiation Warning All areas within the F.1R-PML facility are posted .[ with radiation warning signs to meet the requirements-of 10CFR20. j Access to'high and very high radiation areas are labelled and l2 monitored as required by 10CFR20. t ? 31. !l .I

..~. SAFETY-ANALYSIS Ford Nuclear Reactor I Docket 50-2, License R-28 Revision 2 11.2 Facility Internal Surveillance 11.2.1 Floor Smear Surveys j i Smear surveys generally are done weekly during 2 l norma] operation. A normal week's survey consists of more than 100 smears. Discretion and judgement are used in selecting the l locations to smear. 11.2.2 Contamination Clean-up Standards Contamination cleanup guidelines are established 2 by the Health physicist. The guidelines can be waived on the l recommendation of the health physicist with approval of the Reactor Manager or one of the Assistant Reactor Managers. l2 Access to contaminated areas is restricted until the areas have been cleaned up. 11.2.3 Area Radiation Surveys i Neutron and gamma radiation surveys of the FNR-FML facility are conducted routinely. Survey results for the reactor pool floor, beamport floor, and reactor basement are posted at access points. i 11.2.4 Airborne Surveys i 11.2.4.1 Mobile Airborne Particulate (MAP) Monitor l The pool floor, beamport floor, and beamport exhaust system are continuously monitored for airborne particulate radioactivity by MAP unit:

t these locations.

A weekly MAP survey is conducted to ens 3 proper operation and

t alarm points.

11.2.4.2 Airborne Tritium Survey i -i An airborne survey is conducted whenever i contaminated heavy water transfers are conducted. -A survey record l is maintained. Bioassay is required if greater than 1 MPC is detected by the air sampling or when deemed necessary by the health physicist. 11.2.4.3 Retention Tank Airborne Survey 'I An airborne survey is conducted prior to personnel entry into retention tanks to ensure the requirements j 10CFR20.1501 and 1502 are met. l2 l i b b I 32 't

SAFETY ANALYSIS. Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 11.2.5 Pool Water Analysis ) A pool water sample is taken biweekly. A gamma l2 l isotopic analysis is conducted within fifteen minutes ftr short lived radior.ctivity and after one or two weeks for long lived j2 j radioactivity. Tritium and gross beta analyses are also performed. } 11.2.6 Hood Airflow Surveys An airflow survey is run semiannually on each l2 FNR-PML laboratory exhaust hood. In order to remain in operation, j cach hood must have adequate exhaust airflow. a 11.2.7 Heavy Water Tank Tritium Analysis Each time that heavy water is transferred, a tritium analysis is conducted on the water remcved from the tank. l2 l 1 Reference to Tech Spec Limit ? Deleted l 1 - l l I i i I i 1 l i I I i 2 i l l t s~. -l i s 33 )

'~~ ^ ~ ~ __s _.m m. ' SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 12. CONDUCT OF OPERATIONS 12.1 Organizational Structure and Qualifications (Figure 12.1) The FNR Reactor Manager is responsible for the safe operation of the FNR. He is responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, including the technical specifications and the regulations. During periods of his i absence, his responsibilities are delegated to the Assistant Reactor Manager. In all matters pertaining to the operation of the plant and technical specifications, the FNR Reactor Manager reports to and is directly responsible to the Director, Michigan Memorial-phoenix project. The minimum qualifications'for the FNR Reactor Manager are a bachelor's degree and at least four years of reactor operating i experience in increasingly responsible positions. Years spent in graduate study may be substituted for operating experience on a ^ one-for-one basis up to a maximum of two years. Within six months af ter being assigned this posit ion, the Reactor Manager must apply for an NRC senior operator license if he does not already hold. one. l A health physicist who is organizationally independent of the FNR operations group is responsible for radiological safety at t the facility. ~ A licensed operator or licensed senior operator pursuant to 10CFR55 must be present at the controls whenever the reactor is in operation. The minimum operating crew is composed of two individuals, at least one of whom will be so licensed. A licensed I senior operator must be present or readily available on call at any time the reactor is in operation. All licensed operators at i the -facility participate in an approved operator requalification l program as a condition of their continued assignment to operator duties. 12.2 Operational Review and Audit A Safety Review Coraittee reviews reactor operations and advises the Director of the Michigan Memorial-Phoenix Project in matters relating to the health and safety of the public and.the safety of facility operations. The Safety Review Committee has j eight members of whom no more than the minority may be from the line organization shown in Figure 12.1 or administratively' report t to anyone in that line organization fe' low the Vice President for j Research. The committee is made up of University staff and j faculty who collectively provide experience in reactor engineering, instrumentation and control systems, radiological safety, and mechanical and electrical systems. i 34

, - -. _. -~ SAFETYcANALYSIS~ Ford Nuclear-Reactor Docket 50-2,'-License R-28: i' Revision 2-Figure 12.1 Grgainization Structure and Qualifications j Regents University of Michigan-t President j i Vice President Vice President and Chief-Research Financlal Officer I l l Director Director Michigan Memorial-Phoenix Project Financial Operations and l Sponsored Program Finanw I Safety Review Director Comittee Occupational and Environmen1.al Saiety l 1 i Director Radiation Safety Service I Iluclear Reactor Health Physicist Laboratory Manager I I l Assistant Manager Assistant Manager I Reactor Operations Research Support Activities -l I Line Function Saiety and i Licensing Function Licensed Operators 35 w p y,

1 SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 i i i l The Safety Review Committee reviews and approves proposed j experiments and tests utilizing the reactor facility which are significantly different from tests and experiments previously j performed at the FNR. In the event of a disagreement between the Committee and the Reactor Manager ovci approval of an experiment, j the matter is referred to the Director, Michigan Memorial-Phoenix Project for resolution. In addition, the Committee reviews j reportable occurrences; reviews and approves proposed amendments. I to the facility license; reviews proposed changes to the facility made pursuant to 10CFR50.59(c); and reviews the audit report provided by the consultant for reactor operations. 8 r A consultant audits reactor operations and reactor operational records annually for compliance with internal rules, i procedures, and regulations and with license provisions including 4 technical specifications; audits existing standard operating { procedures for adequacy and to assure that they achieve their intended purpose in light of any changes since their implementation; and audits plant equipment performance with particular attention to operating anomalies, reportable occurrences, and the steps taken to identify and correct their j causes. 12.3 Operating Procedures t Written procedures, including applicable check lists, j reviewed and approved by the Safety Review Committee are in effect and followed for the following operations: startup, operation and i shutdown of the reactor; installation and removal of fuel elements, control rods, experiments and experimental facilities; actions to be taken to correct "ecific and foreseen potential mal functions of systems or components, including responses to alarms, suspected primary coolant system leaks, and abnormal reactivity changes; emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, reentry, recover", and medical support; maintenance procedures which could have an effect on reactor safety; periodic 2 surveillance of reactor instrumentation and safety systems, aren monitors, and continuous air monitors; facility security; and radiation protection. Substantive changes to the above procedures are made only [ with the approval of the Safey heview Committee. Temporary changes to the procedures.that do not change their original intent may be made with approval of the FNR Reactor Manager or the ' Assistant Reactor Manager. All tempo,r.ary changes to the i procedures are documented and subsequently reviewed by the Safety Review Committee. 3 i I 36 t .~.

SAFETY ANALYSIS Ford Nuclear Reactor -Docket 50-2, License R-28 Revision 2 13. BASES FOR TECHNICAL SPECIFICATIONS 13.1 Power and Flow Limiting Combinations in Forced Convection Limiting combinations of power and flow with forced convection are ussed on preventing the clad temperature from reaching the saturation temperature at any point. Limiting combinations are' determined by straightforward calculations of heat conduction from fuel to the primary coolant and heat transport from the core by the primary coolant. Some calculation-of basic fuel and core parameters are necessary. An-experimental determination of core flux profiles and peaking factors is utilized to calculate peak heat generation rate. 13.1.1 FNR Core Parameters Analysis is performed for a 25 element core consisting of 21 standard 18-plate elements and four control rod, 9-plate elements containing 414 fuel plates. This is the minimum critical mass core. The total heat transfer surface area of the core is: As = 414 (2)(wrlr) (13.1) = 414 (2)(0.21 ft)(2 ft) 345 ft2 = where: As = Active fuel surface area, ft2, Active fuel width, ft; wr = = Active fuel length, ft. lt The core flow area is: Ne we te (13.2) Ac = = 414 (0.128 ft)(0.0098 ft) = 0.884 ft2 where: Ne = Number of coolant channels; we = conl a nt channal width, ft; Coolant channel thickness, ft. te = 37

4 i SAFETY ANALYSIS' Ford Nuclear Reactor i Docket 50-2, License R-28 i Revision 2 Calculation of coolant velocity is necessary to f ultimately determine the convective heat transfer coefficient. For this calculation, a minimum flow rate of 900 gpm is used. i F/Ac (13.3)- v = ( 900g/m)(60 m/hr)( f t3/7.49g)/0.884 ft2 = = 8160 ft/hr where: Primary coolant velocity, ft/hr; v = primary coolant flow rate, gpm. F = 13.1.2 Heat Transfer Coefficient The average coolant temperature in the core is 1 assumed to be 170 F for evaluating water preparties._ That value is calculated from a maximum inlet temperature cf 116 F and outlet ~! temperature of 224 F during limiting power-flow cperation. I Reynolds number is calculated utill aing water { properties at 170 F. The characteristic channel aimension, ', based on analysis of flow between closely spaced parallel platcm i is the hydraulic diameter. Hydraulic diameter is equal to four times the cross sectional area divided by the wetted perimeter. vpl/p (13.4) Re = (8160 ft/hr)(60.79 lbm/ft3)(0.0196 ft)/ = (0.90 lbm/ft hr) 10,803 = where: Reynolds number; j Re : primary coolant density, lbm/ft3; I p = Characteristic channel dimension, 2tc; -l 1 = primary coolant viscosity 7"lbm/ft hr. I p = For a smooth channel of assumed relative roughness, E/D.= 5x10-5, the Reynolds number describes flow in the transition region from laminar to turbulent flow on a Moody friction chart, Figure 13.1. -i 38 l

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 Figure 13.1 Friction Coefficient as a Function of Reynolds Number for Various Relative Roughness Ratios 0 100 .,minar Cnbcal Transition l ll ll l l ll ll ll l l og no. zo, IT% Canpi te Turbulena. I'@ ap O 0M0 I-wi 6 4 J 77Tn yg 0,05 0 070 ) ]N-Ibiii I l ii j 0 04 l O twiu ON ] } l l \\ [@A l l l "k s .g 5 / I 0,010 i 'l 5 .' 0am i j ij'j," 6g c t i. f, 0Om y-f O.000 5 - 0.004 i g \\ .l 8 N j 00s d k xt l , 0.002 g c q g A i P 3 A m 00m ( OCCI ~E Smooth Mpes V J ,i DDX2 s X 3 jl g) [*- j l l< OMFJ6 r ~ Riveted Steel D E* 003 ~~ '~ 0.0002 C*nent* 0.001- 0.01 C% Cahamued Iron 0.0006 k l* hN' Cast Imn 0RX25 0.0003 N s. 1 \\ 0 C18 Conunercial Stee.1 or Wrought ima C 00016 0M.05 N-K 0 009 Dn=n hbing 0.000005 QQy 000e I I l iIf1! I I I itf.. # m30 ) 4 s tf giv3.. 3p sg,np ,g J 0.000,01 to' *19) 4 s 10* a:0's 4 s 6 bynotda Nunter 4/D=0*o.00

/D-cac.cos A.

39

i SAFETY ANALYSIS Ford Nuclear' Reactor Docket 50-2, License R-28 I Revision 2. l Prandtl-number is: i pc/k (13.5) Pr = (0.90 lbm/ft br)(1.00 Btu /lbm F)/ = (0.386 Btu /hr ft F) = 2.33 .l where: ~ Prandtl number;. f Pr = Primary c'oolant specific heat, Btu /lbm F; c = Primary coolant thermal conductivity, -f k = Btu /hr ft F. Nusselt number, which is a function of Reynolds and l Prandtl number, is determined using the Dittus-Boelter equation characteristic of transition and turbulent flow. i Nu = 0.023 (Re)0 8 (Pr)0 4 (13.6). j = 0.023 (10,803 )0 8 ( P. 33 )D 4 54.39 l = where: Nu = Nusselt number. The convective heat transfer coefficient for the film at the fuel plate clad-coolant interface.is: r h = k Nu/1 ' ( 13. 7 )- (0.386 Btu /hr.ft F)(54.~39)/(O'.0196 ft) = = 1071 Btu /hr ft2 y 'This value is characteristic of subcooled water flow at near atmospheric pressure. The convect'ive heat transfer coefficient would remain near 1000 for any. coolant. temperatures between 70 F and 212'F and for,all pFa~ctical flow rates. 'I i l J 40. I

SAFETY-ANALYSIS l Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2-For heat conduction from the fuel plate centerline f to the coolant, the thermal conductivities and thicknesses of-the fuel meat and clad are needed. The clad thermal conductivity, Ec, is.115 Btu /hr ft F, the value for 1100 aluminum. The meat thermal conductivity, K, is an experimentally determined value for oxide i cermets similar to powder metallurgy aluminides and oxides that l are authorized for use in FNR fuel equal to 103 Btu /hr ft F. The vclue of Km is close to Ec because the fuel meat is greater the 90 l atom % aluminum. l The thermal resistances of the meat, clad, and convective film are. R. = X /km (13.8) i (.00125 ft)/(103 Btu /hr ft F) = -3

1. 21x 10- 5 hr ft2 F/ Btu

= Xc /kc Re = (.00125 ft)/(115 Btu /hr ft F) (13.9)- = ? 1.09x10-5 hr f t2 F/ Btu = a t 1/h (13.10) Er = 1/(1071 Btu /hr ft2 p) = }

9. 3 4 x10- 4 hr ft2 F/ Btu.

= Rtotal = Hm + Rc + Hr (13.11) l ^

1. 21x 10- 5 ~+

1. 0 9x10- 5 + 9. 3 4 x 10- 4 j

= l 9.57x10-4 hr ft2 F/ Btu = = Rt j t where: = Fuel meat thermal resistance, hr ft2 F/ Btu; R. Clad thermal resistance, hr ft2 F/ Btu; He = Rr : Convective film thermal re_sistance, hr ft2 F/ Btu; t 41 l

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2,. License R-28 .j -Revision 2 Total thermal resistance, hr ft2 F/ Btu; Rtotal = X. = Half meat thickness, ft; Xe = Clad' thickness, ft. I It can be seen that Re and R. are about 1% of Rt. Therefore, the temperature across the thin fuel plate.will be nearly constant. Almost the entire temperature rise that a provides the driving force for heat transfer.from the fuel plate to the primary coolant occurs across the convective film. The criterion for power-flow limiting combinations is that the clad temperature, Te, not. exceed the saturation i 235 F, at a minimum pool; depth of 20 ft to the l temperature, Tsat = bottom of the core. Heat transfer between the clad and the coolant bulk is: (Te-Te)/Rt (13 12) Op x /A. = i where: l Peak heat generation rate, Btu /hr; Qpk = i Te = clad temperature, F; Primary bulk temperature, F. Ts = a 13.1.3 Peak-to-Average Power Ratio j An experimental flux measurement program was performed in October, 1977, to develop analytical descriptions of l the FNR radial and axial flux profiles for the purpose of. ( determining peak-to-average power ratios within the core. l A self-powered rhodium detector was used to measure 17 thermal flux throughout a. core configured from-33 standard'and 6 control elements. Flux was measured at the horizontal midplane of I each accessible element. Flux levels at five axial positions were -measured in central and peripheral elements. All measurements j were-taken after at least 100 Mw-br of operation to assure equilibrium xenon. The following analytical descriptions of thermal neutron flux closely approximate the physical measurements taken. l Mrd [ l I i L 42 .- - -. ~

I SAFETY ANALYSIS Ford. Nuclear Reactor s Docket 50-2, License R-28 Revision 2 l Coordinate Core (in) Direction Analytical Flux Description i x North-South o = omax cos(nx/27 + n/18) y East-West o = o.a x cos(ny/33 ) z Axial o = oma x cos(nz/36) Flux distributions are symmetric in the east-west and axial directions. In the north-south direction, the flux. peak is shifted toward the heavy water tank located at the north-core i face. Peak-to-average flux ratios are determined by expressing flux at any point in terms of peak flux, j i o = opk cos(nx/27 + n/18)cos(ny/33)cos(nz/36) L(13.13) i i where: t o = core thermal flux at coordinate (x,y,z), n/cm2:sec; opk = Peak thermal flux, n/cm2 see; x = North-south coordinate from core center, in; East-west coordinate from core center, in; y = z = Axial coordinate from core center, in. Equation (3.13) is integrated over the core volume and divided by the core volume to determine the average core thermal flux. Average flux is divided into opt to obtain the peak-to-average flux ratio. This calculation is done in each l coordinate direction to determine radial and axial peaking i factors. Core Peak-to-Average l Direction Flux Ratio t North-South 1.23 0 East-West 1.26 -j Axial 1.21 n. Total (TPF) 1.87 } 43 1

SAFETY ANALYSIS i Ford Nuclear Reactor l Docket ",0-2, License R-28' I Revision 2 Since the' reaction rate, Ito, between thermal flux and fuel produces power, these same values are the peak-to-average [ power ratios. Qpk =.TPF Qav (13.14) where: 3 Average core heat generation rate, Btu /hr. Qav = 13.1.4 Limiting Power-Flow Combinations I Overall heat transport from the core via the-primary coolant is: Qav = me ( To -Ti ) (13.15) j l where: m = Primary coolant mass flow rate, lbm/hr; Primary coolant specific heat, Btu /lbm F; c = To = Coolant outlet temperature, F; Ti = Coolant inlet temperature, F. l Equation (13.12) can be solved for clad temperature, [ Tc. i Te = Ts + Qak Rr/As (13.16) The maximum value of Te occurs at the point of the worst combination of coolant bulk temperature and heat generation rate. Maximum bulk temperature occurs at the core outlet where Ta = To. Peak heat generation rate occurs near the axial midcore position. l Because of the difficulty in precisely locating the worst-combination point for Ta and Qpk, it will be conservatively assumed that Qpk occurs where Ts = To. Equation _(13.16) becomes: Tc = To + Qpk Rr /An (13.17) Equation (13.12) then becomes: l (Te - To )/Rt (13.18) l Qph/As = 1 In terms of ' Qa v, equatfFn (13.18) is: ( Tc - To ) /Rr (13.19) Qav TPF/As = 44 1

~ i -SAFETY ANALYSIS-Ford-Nuclear Reactor Docket 50-2, License-R-28 j Revision 2 ?l Solve for Qsv in equation-(13.19) and set-that equal to equation (13.15). l As (Tc - To )/TPF Rr = me (To - T1) (13.20) Solve equation (13.20) for To. l t 'f To = ( As /( As +'mc TPF Rt))Tc + f (mc TPF Hr/(As + me TPF Rt))TA (13.21) Substitute equation (13.21) into (13.15), solve for Qsv in terms of Tc and T1, and express Qsv in terms of reactor power. t I P= (mc As/3.41x106(As + me TPF Rt))(Te-TI) (13.22) .i where: .{ i P = Reactor power, Mw; lj = Qa v /( 3.41x108 Btu /hr/Mw) i As = 345 ft2; f l i c = 1.00 Btu /lbm F; e TPF = 1.87; l

9. 34x10- 4 hr f t2 F/ Btu; Rt =

235 F; Te = 116 F. Ti = Parameter values can be substituted into equations j (13.21) and (13.22) to produce expressions for power and outlet i temperature as functions of core flow rate. l P= 6 gpm/(345 + 0.9 gpm) (13.23)- .i (81,075 + 104 gpm)/(345 + 0.9 gpm) (13.24) To = 1 where.: l l gpm.= Mass flow rate,. gpm;. b/(8.34 lbm/g)(60 min /hr). = 45 jo ?

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 Substitution of various primary coolant flow rates into equations (13.23) and (13.24) yields the following results: primary Coolant Reactor Core Outlet Flow Rate Power Temperature (epm) (Mw) (F) 164 2.00 199 500 3.77 167 600 4.07 162 700 4.31 158 800 4.51 154 900 4.68 151 1000 4.82 149 As mass flow rate increases, the heat capacity of the coolant flowing through the core increases so more power can be generated without exceeding the constraint of 235 F clad temperature. Note that core outlet temperature decreases as flow ~ rate and power increase. From equation (13.18), for a fixed maximum clad temperature, outlet temperature is constrained to decrease at higher powers in order to provide the differential temperature necessary to drive heat from the clad into the coolant. During normal 2 Mw operation of the reactor at a maximum flow rate of 900-gpm, equation (13.15) yields a core outlet temperature of-131 F and a core differential temperature of 15 F. Equation (13.18) can be solved for~ peak clad temperature. Te = To + Qpk Rt/As (13.25) Substituion of appropriate values for Qpk, Rr, and As -produces a. peak clad temperature during normal operation of approximately 166 F. In conclusion, the reactor can be safely operated at a power level of 2 Mw and a-primary coolant flow rate as. low as. 164 gpm without_ exceeding the saturation temperature anywhere'on the fuel. plate cladding surfaces.- AC~~a flow rate of 900 gpm, the reactor can safely. operate up to a power level of-4.68'Mw without exceeding the' saturation temperature anywhere.on the fuel-plate cladding surfaces. 46-

i I SAFETY ANALYSIS-i Ford Nuclear Reactor Docket 50-2, License R-28 'j Revision 2 13.2 Power Limits in Natural Convection In order to determine reactor power limits in natural j convection, a series of core inlet and outlet temperature-measurements were made with the reactor operating below 100 kw in natural convection. The following data summarize the measurements. r i Reactor Primary Coolant Temperature power Core Inlet Core Outlet Delta T (kw) (F) (F) (F) { 4 25.5 101 107.8 6.8 l 'I 51.0 101 109.7 8.7-l 76.5 101 112.8 11.8 Projected -i i 100 16.0 } 380 102.0 i 6 Temperatures were measured with a thermocouple probe. { Twelve outlet measurements were made along'the north-south and j east-west axes of the core, and average' outlet temperature was i calculated from those twelve measurements. Inlet temperature was-l measured beneath the core. That measurement was identical to the bulk pool temperature observed on the' pool temperature monitoring l system. i A graphical projection of core differential temperature versus power level to 100 kw and 380 kw resulted in core differential temperatures of 16 F and 102 F respectively. i The maximum actual core inlet temperature. permitted with the reactor in operation is 131 F. That temperature is the 129 F automatic rod rundown point plus a 2 F measurement error. At a l power level of 100'kw, the core outlet temperature is projected to l be 147 F (131 F+ 16 F), which~is the maximum coolant temperature. j I 1 ) e 47 --- + ~ --- -.- -.m,. l

SAFETY' ANALYSIS Ford Nuclear Reactor-Docket 50-2, License R-28 Revision 2. L Using the projected data, at a power level of 380 kw, the core outlet temperature would be 233 F (131 F + 102 F), which'is the saturation temperature at a pool depth of 18 ft. Minimum permitted pool height above the top of the core is 18 ft. The hottest primary temperature would be at the top of the core; the core outlet in natural convection. In conclusion, operation of the reactor in natural . convection at 100 kw would produce a maximum primary coolant temperature of 147 F if core inlet temperatuare were 131 F. The saturation temperature would be reached at the core outlet if the reactor were operated at approximately 380 kw. 13.3 control Rod Reactivity At the FNR, prompt critical reactivity is approximately 0.0075 delta K/K. By limiting control rod worth to 0.006 delta K/K, the reactor cannot be brought from critical to prompt critical with the control rod. 13.4 Core Excess Reactivity Core excess reactivity is the reactivity necessary to overcome negative reactivity due to: (1) primary coolant temperature increase.from minimum (90 F) to maximum (116 F) operating limits; (2) fission product xenon and samarium buildup in a clean core; (3) power defect due to increasing from a zero j power, cold core to a 2 Mw, hot core; (4) fuel burnup during i sustained operation for 30 days; and (5) moveable experiments. j primary Coolant Temperature Increase i pr = at delta T I (-6x10-s delta K/K/F)(116-90 F) = i -0.0016 delta K/K = i s h. 48 1

i 1 SAFETY ANALYSIS j Ford Nuclear Reactor Docket 50-2, License'R-28 Revision 2' l i Xenon -) i px = -0.022 delta K/K j Samarium _t pa = -0.0075 delta K/K l1 power Defect pro =.-0.0025 delta K/K ] Fuel Burnup (30d)(-0.0003 delta K/K/d) pau = i = -0.009 delta K/K Moveable Experiments j pe = -0.0012 delta K/K l t TOTAL _i i ptotal = pr + px + ps + pro + peu + pe = -0.0436 delta K/K l1 ~ The values of temperature coefficient, xenon reactivity, power defect, and fuel burnup rate are measured.- Samarium i reactivity is calculated based on fission yield and the 1 reactor physics constants associated with the core.. 13.5 Experiment Reactivity j ii The reactivity of secured experiments and total experiments in the FNR core is limited to 0.012 delta K/K based on the i positive reactivity addition values from the Borax' Report l (ANL-5211). The instantaneous addition of 0.016_ delta K/K to the-( FNR core would not.cause any portion of the core lattice to melt. Moveable experiment reactivity limits are designed-to limit 'l the prompt jump of an operating 2 Mw core to the scram setpoint'of i 2.4 Mw and-to produce controllable periods as a result of ( instantaneous insertion or withdrawal' l ,i prompt Jump ( pJ ) ' = De r r ( 1-p )/ ( ETr r -p ) (13.26) { l i i ~ I' 49

-l i SAFETY ANALYSIS l Ford Nuclear Reactor j Docket'50-2', License R-28 ) Revision 2- .j i l -1 Solve equation (13.26) for reactivity. p= Dert (PJ-1)/(PJ-Derr) (13.27) l l 0.0075 (1.2-1)/(1.2-0.0075) = .. g = 0.00126 delta K/K where: l 'I p = Moveable' experiment reactivity, delta K/K; Derr = Delayed neutron fraction, 0.0075; l r i Prompt jump; PJ = 2.4 Mw/2.0 Mw; [ = i 1.2. = 1 t Following a reactivity addition of 0.00126 delta K/K, the reactor would increase in power on a stable period. l T= (Dett-p)/Lp (13.28) l ,i (0.0075 - 0.00126)/(0.1)(0.00126) = = 49.52 sec j i where: ,6 I T = Reactor period, see; I L = Delayed neutron decay constant, sec-1 ; i c ^! A moveable experiment reactivity limit of'O.0012 delta K/K-f ensures that the prompt jump produced by instantaneous insertion j or withdrawal of an experiment will not cause the reactor to exceed the 2.4 Mw limiting safety system setting-(LSSS) and that j the resultant stable periods is longer than the operational limit of 30 seconds. I a l 1 >;)

J i

50 l

. SAFETY ANALYSIS Ford Nuclear Reactor Docket.50-2, License R-28 i Revision 2 I i l l 14. SAFETY ANALYSIS 14.1 Excess Reactivity Addition i 7 The fuel elements used in the Ford Nuclear Reactor are of the same type as those used in the Oak Ridge Bulk Shielding Facility and the Materials Testing Reactor. The transient i behavior of water-moderated lattices utilizing inis type of fuel element has been the object of intensive study in the Borax and Sport tests. The control fuel elements, in which the control and } safety rods travel, have been reinforced to prevent collapse of- -l the safety and regulating rod guide channels in case of a. nuclear i excursion of the type intentionally promoted in the Borax and Sport tests. These control fuel elements are designed to take a 70-psi differential pressure across the guide channels. The data of the Borax Report (ANL 5211) indicates that the I instantaneous addition of about.018 delta K/K.to the barely I critical FNR would melt the. ment of the hottest fuel element if the ambient coolant-moderator temperature were as low as 70 F. The instantaneous addition of 0.016 delta K/K to the reactor I is estimated to cause the center of the hottest fuel plate to rise to about 900 F, some 300 degrees below the melting point, and would result in the release of 34 megawatt-seconds of energy for i an initial coolant-moderator temperatare of 70 F. If this' energy [ were absorbed in the pool water to generate steam, it would give rise to a pressure increase in the building of 1.3 inches of water. The building, as limited by the plumbing traps, is capable of containing a 2-inch increase without a breach of integrity. No i significant release of radioactivity would occur under.these conditions. It is therefore concluded that, if no experiment or' group of-experiments subject to instantaneous removal from the neutron flux { affects reactivity by more than 0.016 delta K/K, then no nuclear accident can cause any portion of the active' lattice to melt. l 14.2 Abnormal Less of Coolant t l Three criteria were used to define an abnormal loss of l coolant event. 1. Fewer than two isolation valves exist between the I reactor pool and a leak source. ? 1 l l 51 ) J

~7 SAFETY ANALYSIS ') Ford Nuclear: Reactor .i Docket 50-2, License R-28 j Revision:2 i i i i 2. Left unabated, the leak would cause the pool water level-I to drop below the top of the reactor core. i 3. Factors-such as thermal stress, high pressure, and the-I potential for mechanical damage from other. equipment, -l cranes, forklift trucks, and the like must exist to make the event credible. l 14.2.1 pool Walls { The reactor pool walls at the level of the reactor core are barytes concrete, six feet-six inches thick. Small i cracks in the concrete which result in slow seepage because of low l temperature thermal cycling of the pool water and the walls during j alternating two megawatt reactor' operation and maintenance shutdowns. A catastrophic failure of the concrete is not possible. The pool walls have shown no evidence of deterioration j after 2ver twenty years of operation. 14.2.2 Thermal Column The thermal column is a six feet. square penetration-in one pool wall. The column is covered by a five-eighths -inch { thick aluminum plate which rests on a three and one-half-inch lip i around its edge. The aluminum plate is backed by approximately j four inches of lead and three feet of graphite. The plate is~ sealed on the pool side and the entire column is~ sealed-and l caulked from the outside. A catastraphic failure of the thermal column is not possible, though a small amount of routine seepage-is experienced. 14.2.3 piping Systems: primary Coolant,' Demineralizer, Coolant Sampling, pool Drain, and Emergency Fill Lines All piping systems associated with the primary coolant system are heavy duty schedule forty steel. All of these systems are located in the reactor basement-which is a low l ceilinged room that has no mechanical handling equipment such as - 1 . cranes and chain hoists and that is not accessible to motorized equipment such as forklift trucks. 'I In addition to the systems not being subject to r :hanical damage, the entire reactor coolant system is at lov to.nperature and pressure. Normal maximum coolant-temperature-is i approximately 120 F. At a temperature of 129 F, the' reactor shuts I down automatically. - The highest pressure in the system at the lowest elevation, 32.7 feet below the-pool surface, 'i s 14.18 psi. } i' i j 52 1 -~ I

i SAFETY ANALYSIS. j Ford Nuclear, Reactor l Docket 50-2, License R-28 Revision 2 i i i 14.2.4 Beamports There are eight six-inch diameter and two eight-inch diameter aluminum beamports labeled A -J. The beamports penetrate 3 the pool wall in a staggered arrangement at four different heights-and terminate on the reactor's heavy water tank as shown-in Figure 3 9.1. The four heights at beamport centerlines when referenced to 1 the bottom of active fuel in the core are 0.38, 1.0,-1.61, and-l 2 28 feet. If the pool level reached the lowest elevation in the .l lowest beamport, active fuel would remain immersed in [ approximately two inches of water and the bottom of the core would be immersed in approximately three and one-half inches. l Beamports are configured as in Figure 14.1. Four barriers to loss of coolant can be provided; not all beamports l have all four barriers. The first barrier is the beamtube itself. A collimator is sealed into the beamtube for the purpose of l reducing an extracted neutron beam to a desired cross sectional area. The collimator can have watertight barriers at the pool end and the outer end. The beamport shield door serves as the fourth l barrier, but.the door is raised when a beamport'is in use which .[ eliminates its utility. At the Ford Nuclear Reactor,.every beamport is j configured with at least two of the barriers labeled one, two, and l three in Figure 14.1, except I beamport. I beamport has a i twenty-nine inch long collimator which is open at both ends.: The outer aperture is a one-half inch by one and one-fourth inch [ rectangle. t In all of the beamports, the collimators do not extend beyond the concrete pool wall. Even if the beamtube were -sheared off, the collimator would remain intact to prohibit loss 1 of coolant, with the exception of I beamport. I beamport is shown diagramatically in Figure 14.2. The centerline of the beamport is 1.61 feet above the bottom of i active fuel in the core. If the pool level drained to just below this centerline, active fuel would remain immersed in l approximately eighteen inches of water. 14.2.5 pneumatic Tubes j! A bundle of eight, one and 7/16 inch diameter, pneumatically operated, aluminum, sample irradiation tubes penetrate the pool floor and terminate adjacent.to the west face of the reactor core. A typical set 6Y two tubes can be seen in Figure 9.1 on'the left side of the core. t i i l 53 I s

~, SAFETY ANALYSIS: Ford Nuclear Reactor-i Docket 50-2, License.R-28 Revision 2 l -) i Figure 14.1 Beamport Cor. figuration 1 ? I l 4 I N e e e S 4 e 9 m O O 9 ee e O e e 8 O O 4 e e e e m. e g G e O 4 e e e c e e e e poor-e a e e e a e O O g t =

Collimator O4 gf

\\ \\ \\ @ ;\\\\\\\\ \\\\\\\\\\ x s s i i Tk Beamtube = e e O 9 4 e e e e

e e

O +

g 5

g 3 e, 3 t h 'f T A-( r e x 54 i

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R. Revision 2 Figure 14.2 Ford Nuclear Reactor Pool Elevation (feet above or <below> bottom of active fuel) 21.17 Pool Level Scram I, y Rsactor Core Active Fuel / 2*17 2.00 9 .9 . Collimator. l v s 1.61 [ -[ ~ 1.36 ..l I Beamport s

Peference (bot t.om 0

of active fuel)

Fool Wall c.17>

c4.23) . ; s. 2,,. <6.77> ~ Pneumatic Tubes -w o e CD i

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 Revision 2 A tube or some tubes could be damaged during experiment handling, fuel handling, and reactor maintenance. Damage could occur such that the pool level.could drop to a final -i level that remained above the top of the core or to a level below l the bottom of the core. It is credible to consider damage 1 equivalent to one entire tube being severed below the bottom of the core. A crimping tool and damage control plugs are in. f place in the reactor basement to stem leakage resulting from such j an event, but analysis of the consequences will assume that flow i I is unimpeded. 14.2.6 Pool Drain Time and Flow Rate Calculations Bernoulli's equation with no friction is b = v2 /2g (14.1) I where: i h = Height of pool surface above rupture, ft; i Flow velocity through rupture opening ft/sec; v = i g = Acceleration of gravity, 32.2 ft/sec2 } Solve equation (14.1) for velocity. v = / 2gh (14.2) I Calculate the flow rate through the rupture. { Substitute equation (14.2) for velocity. l .? w = Av i A/ 2 gh (14.3) { = where: i Flow rate, ft3/sec; i w = A = Rupture cross sectional area, f tz, i Convert flow rate to gallons per minutes. j i 449 A / 2 gh ( 14. 4 ) { ~~ w = i l t 56 l i . -~

^ SAFETY-ANALYSIS Ford Nuclear Reactor-Docket 50-2, License R Revision 2 . Flow rate through a rupture equals the rate at which the pool level drops times the' surface area of the pool. w = -As dh/dt (14.5) where: 1 l As = pool surface area, 210 ft2, j Substitute w from equation (14.3) into equation (14.5) and separate variables. t dt = -( As /A / 2 gh) dh .( 14. 6 ). To calculate the drain time from an initial pool water level down to a lower' level, integrate equation (14.6) between the two levels. l (/ 2 As/A / g) (/ ho - / h) '(14.7) t = where: t = Drain time, sec; 'j t Initial pool level above rupture, ft; j ho = F h = Final pool level above rupture,.ft; f Convert drain time to hours. (.0145/A) ( / ho - / h) (14.8) t = Calculations for flow rate, equation (14.4 ), and drain _ time, equation (14.8), for I beamport and one pneumatic tube are summarized in Table 14.1. These two parameters are calculated: (1) shortly after rupture has occurred when the pool i level has decreased one foot below normal to the level scram i setpoint; (2) when the pool level reaches the top of the core; j and, (3) when the pool level reaches the bottom of the' core. In a the case of I beamport, level will not recede below the beamport which-is 1,61 feet above the bottom of active fuel and 1.78 feet above the bottom of the core as can be seen in Figure 14.2. -i The maximum flow rate as the result of the_ rupture 1 of a 1-7/16 inch diameter pneumatic tube with the pool level 27.94 j feet above the opening where the drain.ing occurs is 209 gpm.- In 'l this most severe case, with no emergency makeup flow, the enre remains completely covered for 3.03 hours and partially covered for 3.58 hours following an automatic low pool level scram. By the time the core begins to uncover, fission product decay heat j 57 l .a-

I SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2,-License R-28 l Revision 2 -j i Table 14.1 Flow Rates and Drain Times Corresponding to Various Reactor Pool Levels for Ruptures at I Beamport and One Pneumatic Tube 'l Pool Level Flow Rate Drain Time l (ft above rupture) (gpm) (hr) 6 t 1 Henmport Poo.' Level Scram 19.56 69 0 i Level ut Top of_ Core 0.56 12 12.39-Level at Beamport Centerline 0 0 14.91;- l i Enet!mati c Tube i Pool Level Scram 27.94 209 0 -l Level at Top of Core 8.94 119 3'.03 Level at Bottom of Core 6.60 102 3.58. i t t I i ? i t i ) l J e--

)

1 l I t 58

SAFETY ANALYSIS Ford. Nuclear ~ React'or' I Docket 50-2, License R-28 -Revision 2 will have decreased form about seven percent of full power.at shutdown to less than one percent of full power.. i ~ 14.2.7 Emergency Makeup Water Emergency makeup for the reactor pool is provided by I a four inch water main that enters the reactor building. The four inch line reduces to three inches and eventually branches to threeL -j three-inch lines that-supply emergency-pool water. Calculations using the fire protection handbook and accounting for valves, elbows, and tees in the lines show the emergency makeup flow rate-to be approximately 600 gpm. This flow rate is almost four times t greater than the maximum loss of coolant flow rate in Table 14.1. j 14.2.8 Conclusions Abnormal loss of coolant from the Ford Nuclear Reactor pool that could result in partial or total uncovering of [ the core can be caused by a rupture in or damage to I beamtube or-the pneumatic tube system. The maximum loss of coolant flow rate, from a pneumatic tube failure, is 162 gpm. The emergency makeup water system, with a flow rate of approximately *00 gpm, exceeds the loss of coolant flow rate by almost a factw of four. If i emergency makeup water were not utilized, the reactor core would remain completely covered for 3.92 hours subsequent to'a' pool' i level scram at which time fission product heat would have decayed- [ to less than one percent of the two megawatt-normal operating power level. Based upon this analysis, the most severe abnormal [ loss of coolant event at the Ford Nuclear Reactor would not'cause core damage. l 14.3 Failed Experiment j i Limits are placed on the radioactivity content of gaseous, j particulate, and volatile reactor experiments to ensure that the exposure of workers in the restricted area and the general public I will result in doses below 10CFR20 limits in the event of an i experiment failure.

I 14.3.1 Experiment Radioactivity Limits Based Upon Exposure j

to personnel Within the Restricted Area of the FNR l Building l 6 14.3.1.1 Assumptions l I a. A restricted area 10CFR20, 2 Appendix'E, Column 3 Derived Air Concentration (DAC) produces a i a dose of 5 rem / year-for the isotope j involved based upon 2000 hours / year. ~ exposure. { -l 59

.y ] SAFETY ANALYSIS Ford Nuclear Reactor' 'I Docket 50-2,- License R-28 l -Revision 2- .) -i l b. Volatile or-dispersible activity in-a pool experiment'is uniformly' dispersed I in the lower 1/4~of the-pool floor volume. l c. pool floor volume, V, is approximately 58,000ccubic feet or

1. 6x10s cc.

d. personnel require no more than'0.1 l hour or'6 minutes to diagnose the ~! experiment failure, initiate building evacuation, and evacuate - the reactor -building. e. Calculations are based upon single i encapsulation of experiments. The-allowable dose fraction for single -l encapsulation is 0.1 DAC. [2' [ 14.3.1.2 Calculations The concentration of radioactivity, C, permitted in the reactor building air is: i C= ( T / t ) (. 0.1 DAC) pCi/cc '(14.9) = 2000-DAC l2 .i where: i T = 2000 working hours per year; l t = Exposure _ time. 0.1 hr. The total activity, A, of an. experiment l based upon_the concentration and volume into.which it is dispersed is* A = CV/4 (14.10) = 8.0 x 1011 DAC pCi l2 14.3.2 Experiment Radioactivity Limits Dased Upon Exposure to personnel in Unrestricted Areas ^" 14.3.2.1 Assumptions i a. A 10CFR20. Appendix-B, Table 3, l2 i Column 1, Air Effluent Concentration (AEC). produces a dose of 0.1 rem / year based upon continuous exposure. 1 b. Stack-dilution is 400-1 60

~ -i SAFETY ANALYSIS .f -Ford: Nuclear Reactor j Docket 50-2,-License R-28: l Revision 2 .{ .i l c. The maximum rate at which air can be l exhausted from the FNR.following i initiation of building evacuation is 'l 300 cfm' based upon opering the exhaust j duct'for the-hood in-R :m.3103. d. The PML exhaust fan-flow rate is i 11,000 cfm. Room 3203 hood airflow is diluted by this flow.. l e.- Calculations are based upon single encapsulation of experiments. The allowable dose fraction for single = encapsulation is 0.1 AEC. j2 l t f. The unrestricted area is continuously j occupied for 2 hours following the. experiment failure. l g. Volatile or dispersible activity in a pool experiment is uniformly dispersed-in the-lower 1/4'of the pool' floor' l volume. l h. Pool floor volume, V, is approximately: 58,000 cubic feet or 1.6x108.cc. l 14.3.2.2 Calculations j i The allowable ground level concentration of radioactivity, Cs, is: j [(365)(24)/2)(0.1 AEC) pCi/cc (14.12) Cs = The allowable stack concentration, Cs,. based upon a 400 dilution factor is: i 400 Cg (14.13) Co = i - 400 ( (365 )( 24 )/2] (0.1 AEC) pCi/cc [2. The allowable exhaust hood concentration, C, based upon mixing with the remainder of the PML exhaust in the stack is: ~ C= (11,000/300)Cs ( 14.14 )- q (11,000/300)400[(365)(24)/2) (0.1 AEC) l2- = = 6.42x106 AEC pCi/cc l2 61 1 'I i

SAFETY ANALYSIS ' Ford Nuclear Reactor Docket 50-2, License R-28' Revision 2 Assuming no fresh air is introduced into the pool floor and that the radioactivity undergoes no decay, the total. experiment activity, A, allowed would be: A = CV/4 (14.15)- j i 2.57x1015 AEC [2 = s 4 14.3.3 Limits on Single and Double Encapsulation -j Experiments j Comparison of equations (14.10) and (14.15) with the insertion of DAC and AEC values shows that the total l2 t radioactivity for an experiment with single encapsulation is limited by exposure to personnel within the restricted area of the l reactor building. The microcurie content of any experiment is calculated using equation (14.10) and DAC values from 10CFR20 l2 for the isotope involved. I For double encapsulation experiments, the dose fraction is increased from 0.1 DAC to DAC because the double [2 l barrier decreases the probability of experiment failure. Therefore, the total radioactivity content of an experiment with double encapsualtion is limited to ten times the content ofaut t experiment with single encapsulation. in the event that an experiment contains more than one releasable isotope, I[Ai/(As)11mit ] 5 1 (14.16) -l where: i Ai = Actual isotope radioactivity, pCi; ( At )1im i t = Equation (14.10) radioactivity limit. pCi. l 14.3.4 Fissile Material Experiment Activity Limit A review of 10CFR20 limits of fissile materials and their fission products shows that iodine-133 has the most restrictive limit due to its combined biological impact and decay rate. In placing activity limits on fissile material er fueled experiments, the assumption is made that the entire fission yield is iodine-133. 62 ^

l 1 .i SAFETY ANALYSIS 't Ford Nuclear = Reactor Docket 50-2, License R-28 Revision 2 l l l A calculation follows to determine the combination f of fissile material weight, neutron flux, and irradiation time l that will produce an amount of iodine-133 activity equal to the limit in equation (14.10). Specific values for U235 are'provided as an example. 14.3.4.1 Calculations The fission reaction rate, R, is given by: R = Noo (14.17) = 1. 4 8x10- 3 Wo U235 fissions /sec f where: i N = Atoms of fissile material; + WAo /M ; j = i W= Sample weight, mg; Avogadro's number, 6.02x1023 atom / mole; Ao = M = Molecular weight, mg; 235,000 mg for uranium-235; = 1 .I o = Microscopic fission cross section, cm2; j 577x10 2 4 cm2 for uranium-235; = 2 n/cm /see, o = Thermal neutron flux, The total number of fissions, F, in a given irradiation time is: F = Rt (14.18) i = 1.48x10-3 Wot U235 fissions i where: Irradiation time, sec. t = Two fission products result per fission, i so the total number of fission produces produced, FP, is: l FP = 2F (14.19) f 2.96x10-3 Wot U235 fission products. = i 4 63

SAFETY ANALYSIS Ford Nuclear Reactor Docket 50-2, License R-28 r Revision 2 The decay rate, D, of radioactive fission products is proportional to the half-life, T2, of the isotope involved. Since it was assumed that all fission products are j I133, the I133 half-life of 20 hours is used. The value of 20 hours is reasonably representative of the average half-life of all fission products. I D= L FP (14.20) i (In2/T2 ) FP = 2.85x10-8 Wot U235 disintegrations /sec = where: l L= Fissile material decay constant, sec-2 ; In2/T2 ; = Fissile material half-life, see; T2 =

7. 2x 104 see for 1133.

= Experiment radioactivity, A, is obtained by converting the decay rate to microcuries. D/ 3. 7x104 pCi (14.21) A =

7. 7x10- 13 Wot pCi 1133

= The combination of fissile material weight, neutron flux, and irradiation time permitted is determined by setting equations (14.10) and (14.21) equal. For the specific example of U235 and its fission product, 1133, I 7.7x10-13 Wot 8.0x1011 DAC l2 =

1. 04x102 4 DAC l2 Wot =

1.04x1017 (single encapsulation) l2 = 1.04x1016 (double encapsulation) = l2 i uhere DAC = 1x10-7 pCi/cc for I133. l2 i 64 i}}

ML20059M316

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=

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